ML20010C804
| ML20010C804 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 07/28/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Koester G KANSAS GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8108210011 | |
| Download: ML20010C804 (7) | |
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Mr. GleM L. Koester RHartfield, MPA
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9 Vice President - Nuclear WJohnston 5
Kansas Gas and Electric Company WHousten O
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Dear Mr. - Koester:
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Subject:
SNUPPS FSAR - Request for Additional Information ' %
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Materials Engineering D
(4 As a result of our review of your application for operating licenses we find that we need additional information regarding the SNUPPS FSAR. The information required is related to Reactor Vessel Material, Reactor Coolant Pressure Boundary Materials, and Reactor Coolant Pump Flywheel, which are being reviewed by the Component Integrity section of the Materials Engineering Branch. The specific information required is listed in the Enclosure.
To maintain our licensing review st.hedule for the SNUPPS FSAR, we will need responses to the enclosed request by September 7, 1981.
If you cannot meet this date, please inform us within seven days after receipt offthis letter of the date you plan to submic your responses to that we may review our j
schedule for any necessary changes. Please note that advance copies of these 1-questions were provided to the SNUPPS staff on July 21, 1981 at the public meeting with the Reactor System Branch.
Please contact Dr. G. E. Edison, Licensing Project Manager, if you desire any discussion or clarification of the enclosed request.
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Sincerely, I
N demed by bhan L. Teu.w.
ggaggg0ggggg2 Robert L. Tedesco, Assistant Director 3
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A PDR-for Licensing i
Division of Licensing l
Enclosure:
As stated cc: Seenextpagaf
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O Mr. J. K. Bryan Mr. Glenn L. Koester Vice President - Nuclear Vice President - Nuclear Union Electric Company Kansas Gas and Electric Company P. O. Box 149 201 North Market Street St. Louis, Missocri 63166 P. O. Box 208 Wichita, Kansas 67201 cc: Gerald Charnoff, Esq.
Shaw, Pittman, Potts, Dr. Vern Starks Trowbridge & Madden Route 1, Box 863 1800 M Street, N. W.
Ketchikan, Alaska 99901 Washington, D. C.
20036 Mr. William Hansen Kansas City Power & Light Company U. S. Nuclear Regulatory Commission ATTN:
Mr. D. T. McPhee Resident Inspectors Office Vice President - Production RR #1 1330 Baltimore Avenue Steedman, Missouri 65077 Kansas City, Missouri 64141 Ms. Treva Hearn, Assistant General Counsel Mr. Nicholas A. Petrick Missouri Public Service Commission Executive Director, SNUPPS P. O. Box 360 5 Choh Cherry Road Jefferson City, Missouri 65102 Rockville, Maryland 20850 Jay Silberg, Esquire Mr. J. E. Birk Shaw, Pittman, Potts & Trowbridge Assistant to the General Counsel 1800 M Street, N. W.
Union Electric Company Washington, D. C.
20036 St. Louis, Missouri 63166 Mr. D. F. Schnell Kansans for Sensible Energy Manager - Nuclear Engineering P. O. Box 3192 Union Electric Company Wichita, Kansas 67201 P. O. Box 149 St. Louis, Missouri 63166 Ms. Mary Ellen Salava Route 1, Box 56 Mr. Tom Vandel Burlington, Kansas 66839 Resident Inspector / Wolf Creek NPS c/o USNRC Eric A. Eisen, Esq.
P. O. Box 1407 Birch, Horton, Bittner & Monroe Emporia, Kansas 66801 1140 Connecticut Avenue, N. W.
Washington, D. C.
20036 Mr. Michael C. Keener Wolf Creek Project Director State Corporation Commission Ms. Wanda Christy State of Kansas 515 N. 1st Street Fourth Floo,, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612
s ENCLOSURE i
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REQUEST FOR ADDITIONAL INFORMATION WOLF CREEK UNIT i10. 1 I
123.0 MATERIALS ENGINEERING BRANCH--COMPONENT INTEGRITY SECTION 123.MC Identify whether SA-540 Class 1 or 2 material was used for closura J
bciting in reactor coolant pumps.
If SA-540 Class 1 or 2 materials were used i
for closure bolting in reactor coolant pumps, demonstrate the generic I
adequacy of the fracture toughness and demonstrate compliance with i
Paragraph I.C of Appendix G, to 10 CFR Part 50.
i, 123.2WCIndicate whether the individuals performing the fracture toughness i
l tests were qualified 'y trairing and experience and whether their competency o
i was demonstrated in accordance with a written procedure.
If the above infor-mation cannot be provided, state why the information carenct be provided and identit/ why the method used for qualifying individuals is equivalent to those of Paragraph III.B.4 Appendix G, 10 CFR Part 50.
I i
j 123.3WCTo demonstrate compliance with the beltlina material test requirements of Paragraph III.C.2 of Appendix G,10 CFR Part 50:
1 Provide a schematic for the reactor vessel showing all welds, plates a.
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and/or forgings in the beltline.
Welds should be identified by shop i
control number, weld procedure qualification number, the heat of filler 1
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s metal, and type and batch of flux.
Provide the chemical composition for these welds (particularly Cu, P,;and S content).
b.
Indicate the post-wtid heat treatment used in the fabrication of the test
- welds, c.
Indicate the plates used to fabricate the test welds.
d.
Indicate whether the test specimen for the longitudinal seams were removed from excess material and welds in the vessel shell course following completion of the longitudinal weld joint.
I 123.4UCTo derionstrate compliance with the fracture toughness requirements of Paragraph IV.A.1 of Appendix G, 10 CFR Part 50:
a.
Provide the RT f r all RCPB welds which may be limiting for operation NDT of the reactor vessel, b.
Indicate whether there are any RCPB heat-affectea zones which require CVN impact testing per paragraph NB-4335.2 of the 1977 ASME Code.
Provide CVN impact test data fo tnese heat-affected zones which may be limiting for operation of the reactor vessel.
c.
Indicate that there are no ferritic KCPB base metals.other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code.
If there are ferritic RCPB base metals that than in vessels which require a
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fracture toughness testing tt NB-2300 of the ASME Code, provide CVN
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i impact and drop weight data for all materials which will be limiting for l
operation of the reactor vessel.
123.S4CRevise the FSAR to indicate that the conclusions of Westinghouse Topical Report WCAP 9292 is applicable to Callaway Unit 1 SA-533 Grade A, Class 2 steel and SA 508 Class 2a steels.
123.6WCProvide actual pressure-temperature limits for Callaway Unit 1 based upon the limiting fracture toughness of the reactor vessel material and th?
predicted shift in the adjusted reference temperature, RTNDT, resulting from l
radiation damage.
The pressure-temperature limits for the following conditions must be included in the technical specifications when they are i
submitted:
1.
Preservice hydrostatic tests, 2.
Inservice leak and hydrostatic tests, 3.
Heatup and cooldown operatiuns, and 4.
Core operation.
3 123.7WCProvide full CVN impact curves for each weld and plate in the beltline L
region.
Provide the data in tabulated and graphical form.
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123.8WCTo cemonstrate the surveillance capsule program complies with f
Paragraph II.C.3 of Appendic H:
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Provide the withdrawal schedula for each capsule, j
b.
Provide the lead factors for each capsule.
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Indicate the estimated reactor vessel end of life fluence at' the k wall thickness as measured from the ID.
a i
123.94CIdentify the location of each material surveillance capsule and the materials in each capsule.
1 For each base metal and heat-affected zone sucveillance specimen provide a.
the specimen type, the orientation of the specimen relative to the 4
principal rolling direction of the plate, the heat number, the component code number from which the sample was removed, the chemical composition especially the copper (Cu) and phospnorus'(P) contents, the melting j
practice and the heat treatment received by the sample-material, j
b.
For each weld metal surveillance specimen provide the weld identification
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from which the sample was removed, the weld wire type and heat identifi-cation, flux type and lot identification, weld process and heat treatment
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used for fabrication of the weld sample.
1 c.' Provide a sketch which indicates the azimuthal location for cach capsule p
relative to the reactor core.
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3 123.10WCIndicate the normal operating temperature of the flywheels and provide CVN impact and drop weight test date from each flywheel that indicates tne RT of the flywheels are 100 F less than their normal operating temperatures.
NDT d
WC ubmit for revies an inservice i,tispection progrerr for the pump 123.11 S
flywheels which complies with Paragraph C.4 of Safety Guide 14, October 27, f
1971.
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